SARAX程序系统在钠冷快堆瞬态分析中的应用  

Application of SARAX Code System in Transient Analysis of Sodium-cooled Fast Reactor

作　　者：贾晓茜 郑友琦[1] 杜夏楠 何明涛 翟梓安 JIA Xiaoqian;ZHENG Youqi;DU Xianan;HE Mingtao;ZHAI Zian(School of Nuclear Science and Technology,Xi’an Jiaotong University,Xi’an 710049,China;China Nuclear Power Research Institute Co. Ltd. ,Shenzhen 518000,China)

机构地区：[1]西安交通大学核科学与技术学院,陕西西安710049 [2]中广核研究院有限公司,广东深圳518000

出　　处：《原子能科学技术》2019年第7期1195-1201,共7页Atomic Energy Science and Technology

基　　金：国家自然科学基金资助项目(11775170)

摘　　要：无保护事故下的瞬态分析是钠冷快堆安全分析的重要内容。基于OECD/NEA发布的MOX-3600和MET-1000基准题,本文利用SARAX程序系统对不同钠冷快堆进行了瞬态计算,分析了堆内各种反应性反馈效应,并计算了无保护失流(ULOF)事故和无保护超功率运行(UTOP)事故下燃料温度和冷却剂温度的变化。计算结果表明:SARAX程序系统在快堆瞬态分析中可给出合理的参数预测结果;ULOF事故对于钠冷快堆是更为严重的事故瞬态,会导致堆内的钠沸腾进而发生严重事故。The transient analysis of unprotected accident is significant in safety analysis of sodium-cooled fast reactor. Based on MOX-3600 and MET-1000 benchmarks published by OECD/NEA, SARAX code system was applied to do transient calculation for different sodium-cooled fast reactors. Various reactivity feedback effects in the reactor were analyzed, and the changes of fuel temperature and coolant temperature during unprotected loss of flow (ULOF) transient and unprotected transient over power (UTOP) transient were calculated. The results show that SARAX code system can give reasonable results of parameter prediction in transient analysis of fast reactor. ULOF transient is more severe for sodium-cooled fast reactor in that it will cause the boiling of sodium and then cause severe accident.

关 键 词：钠冷快堆 瞬态分析 点堆动力学 

分 类 号：TL327[核科学技术—核技术及应用]